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U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-266/0L-93-02

Docket Nos. 50-266; 50-301 Licenses No. DPR-24; DPR-27

Licensee: Wisconsin Electric Power Company231 West Michigan Street - P379Milwaukee, WI 53201

Facility Name: Point Beach Nuclear Plant

Examination Administered At: Two Rivers, WI

Examination Conducted: Week of September 27, 1993

RIII Examiner: ( dM c[ gi/)c:/ nF. Ehrhardt Date

[ (IIhChief Examiner: / p e e- is/,c/crsM. Bailey Date

Approved By: / *//d/OT.'Burdick, Chief DatsOperator Licensing Section 2

Examination SummaryExamination administered durina the week of September 27. 1993(Report No. 50-266/0L-93-02fDRS1)Written examinations and operating tests were administered to three (3)reactor operator (RO) and three (3) senior reactor operator (SRO) applicants.A retake examination (written portion) was administered to one (1) additionalR0 applicant. An exit meeting was conducted on October 1, 1993, with olantmanagement.Results: Three R0 and three SR0 applicants passed the examinatioris. One R0applicant failed the written examination.

The following is a summary of licensee strengths and weaknesses noted duringperformance of this examination:

Strenaths

Facility support of examination administration (Section 4).e

Weaknesses

Written examination pre-review (Section 4).*

Large number of deficiencies in the applicants' knowledge relative to*

other sites within Region III (see paragraph 3a).

9311190030 931110ADOCK0500g6PDR

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REPORT DETAILS

1. Examiners

M. Bailey, NRC RIII, Chief ExaminerF. Ehrhardt, NRC RIIIT. Guilfoil, Examiner, Sonalysts, Inc.K. Parkinson, Examiner, Sonalysts, Inc.

2. Persons Contacted

Licensee Representatives

+G. Maxfield, Plant Manager+*J. Reisenbuechler, Operations Manager+J. Becka, Rrgulatory Services Manager+F. Flentie, Specialist-Regulatory Services

+*R. Seizert, 1 raining Manager+*A. Morris, Training Coordinator+*T. Vandenbosch, Training Coordii:ator*K. Draska, Training Coordinator*P. Matson, Senior Training Specialist*C. Hill, Senior Training Specialist

U. S. Nuclear Reaulatory Commission (NRC)

+*J. Gadzala, Resident Inspector

+Present at the management exit meeting on October 1, 1993.*Present at the training staff exit meeting on October 1, 1993.

3. Initial License Trainino Prooram Observations

The following information is provided for evaluation by the licensee viatheir SAT based training program. No response is required.

a. Written Examination :1

Strenaths

e None

Weaknesses

The post-examination review of the written examination by*

the NRC identified deficiencies in the applicants'knowledge. A majority (50% or more) of the applicantsfailed to provide the correct response for each particular iknowledge area identified below. |

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e Initial indication of a ruptured PRT.(R0 question 1)

* The effect of excessive waste gas compressor blowdown onthe area radiation monitoring system. (R0 question 4)

e The order of preference of methods used to cooldownthe plant following a SGTR. (R0 question 8)

e The effects of a loss of steam demand on reactor powerand temperature when critical in the IR. (P0 question 13)

* Conditions under which fuel may be moved in containment.(R0 question 18)

* The effect of Na0H on containment hydrogenconcentration. (R0 question 19)

* Annunciator indications following discharge of a battery.(R0 question 20)

* Use of procedures related to safety-related work durationrestrictions - SR0s only. (R0 question 43/SR0question 18)

Cause of a rod control urgent failure alarm when*

recovering a dropped rod - R0s and SR0s. (R0 question49/SR0 question 24)

Inputs to turbine trip to reactor trip permissive - R0s*

and SR0s. (R0 question 53/SR0 question 28)

Effect of a loss of the MFW system on the plant - R0s |eonly. (R0 question 59/SR0 question 34) '

Reason for position of 2SI-825C (2SI-825A and B BypasseA0V) during normal operations - R0s and SR0s. (R0question 60/SR0 question 35)

e Protection systea response to a loop Thot failure low -R0s only. (R0 question 61/SR0 question 36)

Power supplies to rod drive MG sets - R0s only. (R0e

question 72/SR0 question 47)Which radiation detectors spike when ambient temperature*

is high - R0s only. (R0 question 74/SR0 question 49)

Indications of a fuel handling accident in the spent fuelepit - R0s and SR0s. (R0 question 75/SR0 question 50)

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e Immediate actions for continuous rod insertion whilerecovering a dropped rod - R0s only. (R0 question 82/SR0question 57)

e Effect of boron precipitation on fuel cladding heattransfer capability - SR0s only. (R0 question 85/SR0question 60)

e RHR isolation valve position for RHR overpressureprotection - R0s only. (R0 question 96/SR0 question 71)

e Actions for a sodium bisulfate spill. (SRO question 77)

* Circumstances when the DSS may authorize voluntary entryinto an LC0 without MSS approval. (SR0 question 78)

e Personnel who may authorize overtime in excess of NRCguidelines. (SR0 question 70)

e When activation of the Emergency Response Data System(ERDS) is required. (SR0 question 80)

e Actions necessary to mitigate a main feed pumprecirculation valve failure. (SR0 question 85)

Note: This is a knowledge deficiency that was alsoidentified in the 1992 examint an report (50-226/0L-92-

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Plant response to a pressurizer pressure channel inputefailure. (SR0 question 98)

b. Job Performance Measures (JPMs)

Strenaths

e None

Weaknesses

During the administration of prescripted JPM questions thefollowing weaknesses were noted:

The majority of applicants were weak regarding knowledge ofesubtritical multiplication and nuclear instrumentationindications.

The majority of applicants were weak regarding knowledge ofethe design basis for the SI accumulators.

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c. Dynamic Simulator Scenarios

Strenaths

* None

Weaknesses

* During the performance of one scenario set, the crewidentified and discussed the operator actions required tomanually isolate excess letdown in the event containmentisolation was initiated. However, the crew failed to followthrough with these actions upon entering an event thatrequired the initiation of containment isolation.

4. Trainina. Operations. Security. Rad Protection. Other

Strenaths

* The licensees' training staff support during dynamic simulator andJPM examination validation and administration was good. Facilitysupport of examination administration was significantly improvedfrom that of previous examinations.

* The examiners received complete cooperation from security andhealth physics personnel. This expedited entrance to the plantand precluded any unnecessary delays in examinationadministration.

Weaknesses

* The written examination pre-review conducted by the facility wasnot thorough relative to the effort expended. This was reflectedby the number of post-review comments.

Reference material provided to the NRC for examination development*

was adequate with the following exceptions: |

Electrical diagrams were not detailed enough to allow adequate ;verification of bus loads in all instances.

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Cross referencing Lesson Plan and Training Handbook informationwith operating procedures and other plant information wasdifficult due to the lack of an adequate index for each section ofthis training material.

5. Written Examination Review

Licensee representatives reviewed the written examination prior toadministration and appropriate changes were incorporated into theexaminations with the exception of question number 47 on the R0 exam.

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Following examination administration the facility received a copy of the '

R0 and SR0 examinations and answer keys for review. Facility post-examination comments and the NRC resolutions are documented in i

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Enclosure 3.

THE NRC post-examination review of the written examination also resultedin the deletion of R0 question 47/SR0 question 22 and R0 question 99/SR0question 74 from the examination. These questions did not meet theguidelines for written examination questions in accordance with i

NUREG-1021, Rev. 7, Operator Licensing Examiner Standards, and NUREG/BR-0122, Rev. 5, Examiners' Handbook for Developing Operator LicensingWritten Examinations.

Additionally, R0 question 86/SR0 question 61 was found to have twopossible correct answers. Either choices "c" or "d" could beinterpreted as being correct. The answer key has been changed to ,

reflect either "c" or "d" as a correct answer. Credit was given for |

either "c" or "d" when this question was graded.

6. Simulation Facility Observations

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Simulator discrepancies identified during the examination are noted in.

Enclosure 4, " Simulation Facility Report". !

7. Exit Meetina

Exit meetings with the Point Beach Nuclear Plant training and managementstaff were held on October 1,1993. Section 2 of this report is a listof those who attended the meetings. The following items were discussedduring the exit meetings:

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* Strengths and weaknesses noted in this report.

Simulator items as noted in Enclosure 4, " Simulation Facility |*

Report." ;

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ENCLOSURE 3

FACILITY COMMENTS AND NRC RESOLUTION OF COMMENTS

R0 Ouestion 16

The DC electrical system is in a normal lineup when the supply breaker for D11trips. Which of the following is true (assume no operator actions)?

a. 480 Vac main generator exciter breaker 2-41 will not open on overload.

b. 13.8 kV bus H03 automatic lockout will actuate.

c. If a safety injection occurs, the AC supply to battery charger IDY01, ,

trips. !

d. If a safety injection occurs, diesel generator G01 starts but noautomatic field flashing occurs.

Answer,

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d [1.0]Reference |

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TRHB 12.7, Rev. 3 I

OP-11A, Rev. 22MDB 3.2.12K/A 063000K301 (3.7/4.1)

Facility Comment / Recommendationi

The question contains no right answer.|

l The safeguards sequence logic diagram (Attachment 1) shows that the emergency 1

| generator G01 start signal comes from either a Unit 1 or Unit 2 train "A" '

safety injection signal.

The train "A" start signal to diesel generator G01 is via slave relays ISI-10X(2SI-10X) and ISI-14X (2SI-14X) (Attachment 2). These slave relays areactuated by master relay ISIA-A (2SIA-A) for automatic safety injectionsignals and ISIM-A (2SIM-A) for the manual safety injection signal. Both the ,

master and slave relays mentioned require 125 Vdc train "A" safeguards power |

| from Unit 1 (Unit 2) to actuate (Attachment 2).

Unit 1 train "A" safeguards protection cabinets 10156 and 10157 receive powerfrom panel D17, breaker 12. Panel D17 receives power from Dll, breaker 29 l(Attachments 3, 4, 5). 1

Unit 2 train "A" safeguards protection cabinets 2C156 and 2C157 receive powerfrom panel D22, breaker 3. Panel D22 receives power from Dll, breaker 32(Attachments 3, 4, 6). |

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On a safety injection signal from either Unit 1 or Unit 2 following a loss of IiDll, diesel generator G01 would not start thus making the proposed answer

incorrect.

This question should be deleted from the R0 exam based on the fact that it || contains no correct answer. ji

NRC Resolution

Comment accepted. This question was deleted from the R0 exam.1'

R0 Ouestion 43/SR0 Ouestion 18

A control room operator worked from 0000 to 0800 on Monday morning. He wascalled back in for an emergency relief and worked from 1200 to 1800. He thenworked from 0000 to 1200 Tuesday afternoon. Which statement describes thesafety-related work rule that was violated?

a. An individual should not work more than 12 hours straight.

b. There should be a 12 hour break between work periods.

c. An individual should not be permitted to work more than 16 hours in any 1

24 hour period.

d. An individual should have 8 hours off prior to working 12 hours.

Answer

c [1.0]Reference i

PBNP 3.4.4, 4.1, Rev. 8, pg. 2PBNP 4.3, 3.2, Rev. 15, pg. 15K/A 194001A103 (2.5/3.4)

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Answers "c" and "d" are both correct.

PBNP 3.4.4, " Work Duration Restrictions," step 4.1.2, justifies "c" as beingcorrect. However, step 4.1.3 justifies "d" as also being correct. Eighthours off should be allowed between work periods whether it is an 8 hour or 12hour shift.

Both "c" and "d" should be acceptable

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NRC Resolution

Comment accepted. The answer key has been changed to reflect either "c" or"d" as a correct answer. Credit was given for either "c" or "d" when thisquestion was graded.

IR0 Ouestion 47/SRO Ouestion 22

When must state and county agencies be notified of event emergencyclassification?

a. Within 15 minutes of declaration

b. Within 30 minutes of declaration

c. Within 1 hour of declaration

d. Within 4 hours of declaration

Answer

a [1.0]| Reference

EPIP 1.1, step 3b, Rev. 18, pg. 4K/A 194001All6 (3.1/4.4)Facility Comment / Recommendation

The question is at an SR0 knowledge level at Point Beach Nuclear Plant. It isabove the knowledge level expected of an RO per the PBNP job and taskanalysis.

Notification of offsite agencies is the responsibility of the duty shiftsuperintendent and the technical support center. An R0 does not haveresponsibility for notifications in accordance with EPIP 2.1, " State andCounty Agency Notification." At Point Beach Nuclear Plant, theresponsibilities for notifications rests upon the security shiftcommander / designated offsite communicator and the emergency support manager asstated on Page 6 of EPIP 2.1.

Delete from the R0 exam.

NRC Resolution

Comment accepted. At the time of validation this question was believed to bevalid based upon a review of the facility's training material and learningobjectives for an R0. However, the NRC examiners determined upon furtherreview that all four answers could be interpreted as correct. Therefore, this iquestion has been deleted from the R0 and SR0 exams and was not considered inthe grading of either exam. !

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RC Ouestion 77/SR0 Ouestion 52

The fire protection jockey pump just automatically secured following fillingof the accumulator tank to 130 psig when the relief valve on the accumulatortank lifts (setpoint is 150 psig). Air and system pressure drop to 90 psigbefore a local operator reseats the valve. What is the status of the fireprotection system at this time?

Assume that the fire protection system was in a normal lineup prior to therelief lifting.

Jockey Pump Motor Driven Pumo Air Comoressor

a. running stopped stopped

b. running running stopped

c. stopped stopped running

d. stopped running running

Answer

d [1.0)Reference

TRHB 11.14, Rev. 5LP0003, 3.5, Rev. 13

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K/A 086000K402 (3.0/3.4) |

Facility Comment / Recommendation i

Answers "b" and "d" are correct.

The question states that the jockey pump just automatically secured following i

filling of the accumulator tank to 130 psig. For the jockey pump to have jstarted, the "ER" relay would have to be deenergized (see Attachment 1). The '

question states that the pump secured on pressurizing the accumulator to 130psig, thus contacts "MS-B" and "MS-A" are open, while the "ER" contacts are asshown since the "ER" relay is still deenergized (jockey pump secured onpressure, not level). With the "ER" relay deenergized, the air compressor,

' cannot be started and the jockey pump will be enabled to start on a decreasingpressure of 120 psig. The question then states that air and system pressure

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drop to 90 psig which would cause the "MS-B" and "MS-A" contacts to close and|start the jockey pump and the motor driven fire pump would start on a lowsystem pressure of 95 psig as sensed by PS-3716 (see Attachment 2). Under theconditions given, the jockey pump would be running, the motor driven fire pumpwould be running, and the air compressor would be stopped which matches answer"b".

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Because the question did not state the level in the accumulator, based on thecandidate's assumption, "d" may also be correct. The level difference between i

the short and long probe is 11 inches which takes less than one minute tochange using the jockey pump. With ti.- air leak at the top of theaccumulator, the jockey and electric fire pump would be running at 90 psig.By the time the operator takes action, level would be at the short probe.With level at the short probe, the "ER" relay is energized opening the j

contacts in line with "HS-A" and closing the ER contact in line with "R1" and i

"R2". Since the jockey pump was originally running, it wculd now secure dueto relay "R2" energizing and opening the "R2" contact in link with relay "Rl".this would result in the jockey pump being off, the electric fire pump on, andthe air compressor on resulting in "d" being a correct answer.

Because it is not clear in the question where the water level is during thisevolution, either "b" or "d" is correct depending on the candidate'sassumption.

Both "b" and "d" should be acceptable.

NRC Resolution

Comment accepted. However, changes to the answer choices were recommended bythe facility during the examination review and subsequently incorporated intothe examination. The answer key has been changed to reflect either "b" or "d"as a correct answer. Credit was given for either "b" or "d" when thisquestion was graded.

R0 Question 88/SR0 Ouestion 63

What action is required per AOP-9B " LOSS OF COMP 0NENT COOLING" if a VCT hightemperature alarm is received and NO CCW pumps are running?

Ia. isolate letdown '

b. initiate manual makeup to cool the VCT

c. start P33 and recirc the VCT to the spent fuel pool

d. shift suction of charging pumps to RWST

Answer

d [1.0]

Reference

A0P-9B, 7.3.4, Rev. 4

Facility Comment / Recommendation

Answers "a" and "d" are both correct.

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With no CCW pumps running per A0P-98, Step 7.3.1, letdown would be securedmaking "a" the correct answer. Then when the VCT high temperature alarm isreceived, per A0P-9B, Step 7.3.4, the ope-ator would be required to shiftcharging pumps to the RWST making "d" also correct.

Both "a" and "d" should be acceptable.

iNRC Resolution

Comment accepted. The answer key has been changed to reflect either "a" or"d" as a correct answer. Credit was given for either "a" or "d" when thisquestion was graded.

R0 Ouestion 69/SR0 Ouestion 44

The plant is operating normally at 45% power. If reactor coolant Tctransmitter Tc 401B fails high then the steam dump solenoid valves (SV-2050B-20578) for the positioner air signals......

a. reposition to open the dump valves and the dump valves dump steam

b. do not reposition but the dump valves modulate open and dump steam

c. reposition to open the dump valves but the dump valves do not open

d. do not reposition and the dump valves do not open

Answer..

c [1.0] |

Reference;

TRHB 13.9, 2.1, 2.2, Rev. 1LP0253, 3.3.3, Rev. 3K/A 016000K303 (3.0/3.1)

Facility Comment / Recommendation

Answers "c" and "d" are both correct.

The sentence "...the steam dump solenoid valves (SV-2050B-2057B) for the |positioner air signals..." refers to two separate components in the steam dump I

system.

SV-20508 - 2057B are " ARMING" solenoids in series with the air operatedpositioner valves (see Attachment 1). These valves open when a load reductionof 10% within 120 seconds is sensed from turbine first stage pressure(PT-486). Their operation is not affected by auctioneered Tavg. Thesesolenoid valves "do not reposition and the dump valves do not open" upon afailure of Tc 401B making "d" a correct answer. ;

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The oositioner is the valve which controls the modulation of the condensersteam dumps (Attachment 1). With no turbine trip signal present, auctioneered '

Tavg is compared with the reference temperature (based on turbine impulse i

pressure, PT-485) and a modulation signal is sent to the positioner (seeAttachment 2). A failure of Tc 401B high would be sensed as increasing

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auctioneered Tavg and would cause the positioner to " reposition to open the )dump valves but the dump valves do not open" making "c" a correct answer also.

1The way the question is written, some students felt it was asking for theresponse of the positioner, while others felt it was asking for the response ,

of the solenoid valves themselves. This does not represent a misunderstanding I

of how the system operates, but, a misunderstanding of the question. Because jof this, both answers "c" and "d" should be accepted. j

Both "c" and "d" should be acceptable.

NRC Resolution !

I Concur with the facility comment regarding the ambiguous and confusing wording |of the question. Note that the examination proctors received no requests for |

clarification of this question from the applicants. Upon further NRC review, |examiners determined that this question does not meet the guidelines jestablished in NUREG-1021, Rev. 7, Operator Licensing Examiner Standards, and i

NUREG/BR-0122, Rev. 5, Examiners' Handbook for Developing Operator Licensing jWritten Examinations. This question has been deleted from the R0 and SR0 |exams and was not considered in the grading of either exam.

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Enclosure 4

SIMULATION FACILITY REPORT

Facility Licensee: Point Beach Nuclear Plant

Facility Licensee Docket Nos. 50-266; 50-301

Operating Tests Administered: Week of September 27, 1993

The following observations were made by the NRC examination team during thesimulator portion of the September 1993 initial examination. Theseobservations do not constitute audit or inspection findings and are not,without further verification and review, indicative of non-compliance with 10CFR 55.45(b). These observations do not affect NRC certification or approvalof the simulation facility other than to provide information which may be usedin future evaluations. No licensee action is required in response to theseobservations.

! ITEM DESCRIPTION

Simulator computer lock-up During the performance of onescenario a simulator lock-upoccurred which required a fullreboot of the system. Causeunknown.4

11 step differencc between During performance of JPM PB-01,plant computer plasma display Perform Rod Control Exercise With

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and group position (step counter) One Stuck Rod, the plasma displayconsistently indicated 11 steps '

greater than the step counterindications of bank demand for allRCCA groups. Cause unknown.

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